Mechanism Study and Numerical Simulation on Blistering of Plate-Type Fuel

In this paper, the occurrence mechanism of blistering was studied and development processes of blistering were summarized. In addition, a thermal-mechanic-material coupling analysis code, named FROBA-PLATEs (Fuel Rod Behavior Analysis for PLATEs), was developed for plate-type fuel with the consideration of burnup effect. FROBA-PLATEs code was applied to perform the behavior analysis of a dispersion-plate-type fuel. Significant phenomena, including fission gas release and matrix damage, were simulated and key parameters, such as temperature profile, stress and strain profile, were obtained. Most important of all, the starting time of blistering was gained according to the deformation of cladding. The result indicates that: blistering happened at high burnup stage; power density and thickness of cladding are sensitive parameters for blistering. Reducing the power density or enlarge the thickness of cladding can delay or prevent blistering. Furthermore, the influence of blistering on thermal-hydraulic performance was preliminarily investigated by CFD simulation. The simulation result indicates that blistering results in deterioration of heat conduction in the fuel plate.

Thermodynamic Performance Analysis of Megawatts Portable Nuclear Power System Based on Regenerative Brayton Cycle

Nuclear energy is a challenging and ambitious choice for space power system in contrast to solar and chemical fuel. It is able to realize high power and long operating time simultaneously to meet the need of potential applications. Aiming at the thermodynamic performances of the regenerative Brayton cycle with two-stage compression, the paper is objective to get a set of reasonable and competitive operating parameters for the design of the space nuclear power system. Thermodynamic process calculation is applied to analyze the relations of cycle efficiency and influence factors including compression ratio, gas temperature at cold side and hot side, recuperator efficiency, system pressure. The mass estimate model is established to calculate total mass and specific mass of the system with the variation of such design parameters. The calculating results using MATLAB code show that the optimal compression ratio of single compressor varies between 1.2 and 2 along with the other parameters. Either decreasing the cold side temperature or increasing the hot side temperature contributes to enhance the cycle efficiency to about 50%. When the recuperator efficiency changes from 60% to 98%, an ideal heat exchange efficiency, the efficiency corresponding to the optimal compression ratio increase from 35.8% to 52%. But the total mass will also rise from 9.1 tons to 29 tons. It is concluded that the system with cold side and hot side temperature of 450 K and 1300 K, recuperation efficiency of 80% is capable to obtain the maximum cycle efficiency of 36% and the system mass of 10.2 tons. Supposing a space nuclear power system with thermal power of 5 MW, the specific mass is only 5.8 kg/kWe, which indicates obvious technical and economic advantages.

Characterizing the United States Nuclear Used Fuel Inventory Using the Cyclus Fuel Cycle Simulator

The used fuel inventory of the United States commercial nuclear fleet has been accumulating since the inception of nuclear reactors. In order to understand the mass and composition of the used fuel inventory, a nuclear fuel cycle simulation package (Cyclus) is used with a reactor modeling tool (Bright-lite). The parameters for the simulation are obtained as historical operation and burnup data for every reactor in the US fleet, taken from the U.S. Energy Information Administration. The historical burnup data is used to calculate the fuel enrichment of every reactor at every refueling. Discharged uranium inventories calculated by the software are shown to closely match the reference data. The total mass of three major actinide groups are presented as they build up over time. In addition, the evolution of the plutonium composition in discharged fuel is also presented, illustrating Cyclus’ ability to track the composition of material flowing through a large, evolving reactor fleet over decades.

MOCUM Code Verification and Sensibility Study Using C5G7 Benchmark

This work uses the 2-D C5G7 benchmark to verify the accuracy of the MOCUM code, a parallel neutronics program based on the method of characteristics (MOC) for solving arbitrary core geometry. Compared to the MCNP results, MOCUM k-eff, maximum assembly and pin power percentage errors are 0.02%, −0.06%, and 0.64%, respectively. The results demonstrate the high accuracy of the MOCUM code. The calculation uses a total of 56 threads, and the runtime on dual Intel Xeon E5-2699 v3 CPUs is 26 minutes, with speed up higher than 50 times. The sensitivity study of various MOC parameters using the calculation of the C5G7 benchmark problem is also performed. The study reveals that C5G7 requires the usage of 48 or more azimuthal angles. The strong flux gradient and the heterogeneous effects need fine unstructured meshes to resolve. The simulation uses 258 million zones with an average mesh size of 0.016 cm2. The investigation of the polar angle quadrature indicates that Leonard polar angle performs slightly better than Gauss-Legendre and Tabuchi polar angles and more than three polar angles are not necessary. In addition, parameter sensitivity study shows that coarse parameters are prone to introduce error to the neutron flux but not k-eff.

Neutron Up-Scattering Effect in Refined Energy Group Structure

Aiming at generating a 361-group library, this paper investigated neutron up-scattering effect in the 361-group Santamarina-Hfaiedh Energy Mesh (SHEM). Firstly, the Doppler Broadening Rejection Correction (DBRC) method is implemented to consider the neutron up-scattering effect in Monte Carlo (MC) method. Then the MC method is employed to prepare resonance integral table and scattering matrix for afterward calculation. Numerical results show that the neutron up-scattering affects kinf by ∼200 pcm at most for UO2 pin cell problems in the 361-group SHEM, while the fuel temperature coefficient (FTC) is also influenced by 12∼13%. It has also been found that both of the above two influences acts through scattering matrix rather than self-shielded absorption cross sections. In addition, the self-shielding effect of cladding is studied and it’s been found that it affects kinf by 30∼70 pcm.

Analysis and Prospect of the Duplex Fuel Pellets of LOWI Type for Water-Cooled Reactors

The duplex pellets under a “Low-Interact” (LOWI) nuclear fuel design, which consist of an outer enriched annulus and a depleted or natural core, can provide lower center temperature and reduced probability of pellet-clad mechanical interact (PCMI). Analysis and experiments were done in 1970s to examine the benefits and cost of LOWI design for water-cooled reactors. Results showed that the additional economic cost of this design should not be neglected in spite of the benefits. However, due to the improvement of nuclear fuel fabrication technology in the past 30 years, the benefits of LOWI design become more significant, especially when the potential of other methods to elevate the power density and overcome the constraints on ramp rates in power reactors is running out. In order to evaluate the feasibility of deploying the LOWI fuel in commercial and research reactors, neutronics and thermal calculations are made to figure out the performance of duplex UO2 pellets in particular reactors. It is shown that the center temperature of pellet has been greatly reduced without any change on assembly and core geometry, which means the opportunity of less fission gas production, higher power density and more adequate safety margin. A mechanical analysis of a typical LOWI design is also done. The challenges on duplex pellet manufacture are also discussed. Several fabrication techniques are presented to show the potential of cutting the cost of pellet production.

Design of the Spent Fuel Storage Well of HTR-PM

There are more than 400 reactors in operation to generate electricity in the world, most of them are pressurized water reactors and boiling water reactors, which generate great amount of spent fuel every year. The residual heat power of the spent fuel just discharged from the reactor core is high, it is required to store the spent fuel in the spent fuel storage pool at the first 5 years after discharged from the reactor, and then the spent fuel could be moved to the interim storage facility for long term storage, or be moved to the factory for final treatment. In the accident of the Fukushima in 2011, the spent fuel pool ruptured, which led to the loss of coolant accident, it was very danger to the spent fuel assemblies stored in the pool. On the other hand, the spent fuel stored in the dry storage facility was safe in the whole process of earthquake and tsunami, which proved inherent safety of the spent fuel dry storage facility. In china, the High Temperature gas cooled Reactor (HTR) is developing for a long time in support of the government. At the first stage, HTR-10 with 10MW thermal power was designed and constructed in the Institute of Nuclear Energy Technology (INET) of Tsinghua University, and then the High Temperature Reactor-Pebble bed Modules (HTR-PM) is designed to meet the commercial application, which is in constructing process in Shandong Province. HTR has some features of the generation four nuclear power plant, including inherent safety, avoiding nuclear proliferation, could generate high temperature industrial heat, and so on. Spherical fuel elements would be used as fuel in HTR-PM, there are many coating fuel particles separated in the fuel element. As the fuel is different for the HTR and the PWR, the fuel element would be discharged into the appropriate spent fuel canister, and the canister would be stored in the appropriate interim storage facility. As the residual power density is very low for the spent fuel of HTR, the spent fuel canister could be cooled with air ventilation without water cooling process. The advantage of air cooling mode is that it is no need to consider the residual heat removal depravation due to loss of coolant accident, so as to increase the inherent safety of the spent fuel storage system. This paper introduced the design, arrangement and safety characteristics of the spent fuel storage well of HTR-PM. The spent fuel storage wells have enough capacity to hold the total spent fuel canisters for the HTR-PM. The spent fuel storage facility includes several storage wells, cold intake cabin, hot air discharge cabin, heat shield cylinders, well lids and so on. The cold intake cabin links the inlets of all the wells, which would be used to import cold air to every well. The hot air discharge cabin links the outlets of all the wells, which would be used to gather heated air discharged from every well, the heated air would be discharged to the atmosphere through the ventilating pipe at the top of the hot air cabin. The design of the spent fuel storage well and the ventilating pipe could discharge the residual heat of the spent fuel canisters in the storage wells, which could ensure the operating safety of the spent fuel storage system.

Radiation-Induced Electrolytic Corrosion of LWRS: (Part 2) — Verification of In- and Out-Core Redox Potential Differences

The author recently identified that there should exist a “differential radiation cell” mechanism in the reactor water, prompting “radiation-induced electrolytic (RIE)” phenomena. This mechanism was identified while trying to theoretically reconstruct the potential differences observed in two in-pile test loops; NRI-Rez in Czech Republic and INCA Loop in Sweden. Part 2 of this series focuses on the theoretical reconstruction of the observed potential differences. Assuming a state of equilibrium, the author tried to develop a formalism by extending the Nernst equation to reproduce the observed redox potential differences. The radiological potential shift term is separated from the Nernst equation where the latter deals only with stable molecular and ionic species. The radiological effect is described as a perturbation term to the Nernst equation representing a potential shift due to radiation-chemical reactions which should diminish to zero without radiation. The theory generally reproduced the experimental results after fitting the theoretical curve at a single point of the potential for both PWR and BWR-NWC water chemistry environments. This discrepancy is likely due to the “conductive-dielectric property” of the reactor water.

First-Principles Study of Atomic Hydrogen and Oxygen Adsorption on Doped-Iron Nanoclusters

Density functional theory calculations have been used to calculate the ground state structure and oxygen and hydrogen adsorption properties of the pure and doped-iron nanoclusters. Small atomic clusters containing two to six atoms have been considered and a single Fe atom has replaced by a minor element i.e. Zr, Ti, and Sc. Doping of a minor element increases the cluster stability and octahedron Fe5Zr is the most stable structure within this study. Zr- and Sc-doped clusters have the highest oxygen and hydrogen adsorption energy. The electronic structure shows a strong hybridization between the metal 3d and oxygen 2p orbitals with a small contribution from metal 4s and 3p orbitals. Additionally, H s and metal 4s states form a new peak below the Fermi energy and a small modification is observed for 3d orbitals near the Fermi level. A small amount of Zr- and Sc-doping into the Fe-based alloys might improve the oxide film adherence.

Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.